CA1156377A - Neutron activation analysis - Google Patents
Neutron activation analysisInfo
- Publication number
- CA1156377A CA1156377A CA000357743A CA357743A CA1156377A CA 1156377 A CA1156377 A CA 1156377A CA 000357743 A CA000357743 A CA 000357743A CA 357743 A CA357743 A CA 357743A CA 1156377 A CA1156377 A CA 1156377A
- Authority
- CA
- Canada
- Prior art keywords
- sample
- mev
- gamma
- location
- aluminium
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Expired
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Classifications
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01N—INVESTIGATING OR ANALYSING MATERIALS BY DETERMINING THEIR CHEMICAL OR PHYSICAL PROPERTIES
- G01N23/00—Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups G01N3/00 – G01N17/00, G01N21/00 or G01N22/00
- G01N23/22—Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups G01N3/00 – G01N17/00, G01N21/00 or G01N22/00 by measuring secondary emission from the material
- G01N23/221—Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups G01N3/00 – G01N17/00, G01N21/00 or G01N22/00 by measuring secondary emission from the material by activation analysis
- G01N23/222—Investigating or analysing materials by the use of wave or particle radiation, e.g. X-rays or neutrons, not covered by groups G01N3/00 – G01N17/00, G01N21/00 or G01N22/00 by measuring secondary emission from the material by activation analysis using neutron activation analysis [NAA]
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01N—INVESTIGATING OR ANALYSING MATERIALS BY DETERMINING THEIR CHEMICAL OR PHYSICAL PROPERTIES
- G01N2223/00—Investigating materials by wave or particle radiation
- G01N2223/07—Investigating materials by wave or particle radiation secondary emission
- G01N2223/074—Investigating materials by wave or particle radiation secondary emission activation analysis
- G01N2223/0745—Investigating materials by wave or particle radiation secondary emission activation analysis neutron-gamma activation analysis
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01N—INVESTIGATING OR ANALYSING MATERIALS BY DETERMINING THEIR CHEMICAL OR PHYSICAL PROPERTIES
- G01N2223/00—Investigating materials by wave or particle radiation
- G01N2223/60—Specific applications or type of materials
- G01N2223/616—Specific applications or type of materials earth materials
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01N—INVESTIGATING OR ANALYSING MATERIALS BY DETERMINING THEIR CHEMICAL OR PHYSICAL PROPERTIES
- G01N2223/00—Investigating materials by wave or particle radiation
- G01N2223/60—Specific applications or type of materials
- G01N2223/643—Specific applications or type of materials object on conveyor
Abstract
ABSTRACT
Neutron activation is used to analyse the silicon and aluminium content of samples of material, such as bauxite ore and coal. The analysis can be performed on bulk samples, on material on a moving conveyor belt, or on the walls of a borehole. The method involves high energy neutron irradiation of the sample, measurement of the thermal neutron flux in the sample, and monitoring the gamma radiation from the sample at energies of (a) 1.78 MeV and (b) 0.844 MeV and/or 1.015 MeV. Such gamma radiation is produced on decay of (a) 28Al and (b) 27Mg isotopes, produced by the reactions 28Si(n,p)28Al and 27Al(n,p)27Mg. A
single gamma ray detector is used. The analysis preferably utilises equations which include terms to compensate for (a) the production of radiation at 1.78 MeV as a result of the production of the isotope 28Al from 27Al by the thermal neutron reaction 27Al(n,y)28Al, and (b) the Compton scattering of 1.78 MeV gamma radiation and background.
-oOo-
Neutron activation is used to analyse the silicon and aluminium content of samples of material, such as bauxite ore and coal. The analysis can be performed on bulk samples, on material on a moving conveyor belt, or on the walls of a borehole. The method involves high energy neutron irradiation of the sample, measurement of the thermal neutron flux in the sample, and monitoring the gamma radiation from the sample at energies of (a) 1.78 MeV and (b) 0.844 MeV and/or 1.015 MeV. Such gamma radiation is produced on decay of (a) 28Al and (b) 27Mg isotopes, produced by the reactions 28Si(n,p)28Al and 27Al(n,p)27Mg. A
single gamma ray detector is used. The analysis preferably utilises equations which include terms to compensate for (a) the production of radiation at 1.78 MeV as a result of the production of the isotope 28Al from 27Al by the thermal neutron reaction 27Al(n,y)28Al, and (b) the Compton scattering of 1.78 MeV gamma radiation and background.
-oOo-
Description
This invention concerns the simultaneous measurement of the chemical concentrations of the silicon and aluminium constituents of materials. It provides both a method and apparatus for that purpose. There are many possible applications of the present invention, including the measurement of silicon and aluminium in coal and in iron ores. However, the invention was developed primarily to permit the monitoring-of the chemical concentrations of aluminium and silicon in bauxite ores, as part of a quality 1 a control process for the mineral industry, and it is this application of the invention that will be described in detail in this specification.
There are two particular areas in the bauxite industry where the present invention will be used. One is the monitoring of ore quality during ship-loading operations of bauxite for export, where monitoring of aluminium grade and silicon impurity concentrations are essential to ensure that the ore satisfies export contract specifications.
The other use is the monitoring of ore quality whilst sorting the bauxite into stockpiles of different specified chemical concentrations of silicon and aluminium.
Depending on the way in which they have been formed, these stockpiles (a) may contain ore which has been blended , for ore treatment plants or (b~ may be used in subsequent blending operations.
In both of these situations, the current practice in monitoring the ore quality involves the periodic sampling of the ore from the bulk supply, which is usually moving on a conveyor belt when the sample is taken. The samples are moderately large (several kilogrammes) and are either subsampled immediately, or are mixed with other samples, taken by a predetermined number of automatic sampling cycles to form a representative bulk sample, which is then sub-sampled. Sub-sampling and crushing proceeds until a small specimen (of about 1 g) is ultimately available for chemical analysis by wet chemical assaying or by x-ray fluorescence analysis procedures. These sample preparation 115~377 procedures are particularly time-consuming if good representivity of the bulk is required in the sample.
The analysis is also time-consuming.
It has been found that in some situations (for instance during ship loading), when variations of ore quality occur these analytical methods are not fast enough to permit steps to be taken to correct the chemical concentrations of aluminium and silicon (for example, by further blending measures). If it were possible to apply the prior art techniques to on-stream analysis of bauxite on a moving belt, a more rapid analysis of the constituents and hence more rapid corrective blending measures should, in principle, be possible. Unfortunately, wet chemical methods cannot be applied to an on-stream situation, and on-stream x-ray fluorescence methods are inapplicable to lump-flow , measurement. The x-ray fluorescence method is also j unsuitable for the ahalysis of untreated bulk samples, due to the low penetration of x-radiation (less than 1 mm), and the fact that the ore is heterogeneous as regards moisture and particle size.
Neutron activation analysis, which is the basis of the present invention, does not have the problems associated with wet chemical assaying or the x-ray fluorescence technique, noted above, when applied to the analysis of large bulk samples. Indeed, activation analysis methods are directly applicable to large bulk samples and require minimal sample preparation in terms of crushing and drying. They also avoid most of the heterogeneity problems associated with the application of x-ray fluorescence analysis to bulk samples because the neutrons and gamma rays involved have a much deeper pehetration than x-rays. For this reason, the monitoring of bauxite ore ~uality (and the aluminium and silicon content of other materials) on moving belts is amenable to '` `' .:
'. ~ ':
- . .
115~377 the neutron activation method of the present invention.
Neutron activation methods have previously been applied to the analysis of silicon and aluminium in small samples (less than 150 g)-. For example, they ha~e been described in the paper by F. Dugain and J. Tatar in Ann. Inst. Geol. Publici Hungary, Volume 54, p 375 (i970), and in the papers ~y L. Alaerts, J.P. Op de Beeck, and J. Hoste, in Anal. Chim. Acta, Volume 70, p 253 (1974) and in Anal. Chim. Acta, Volume 78, p ~29 (1975). These methods depend on two interactions with the constituent chemical elements. One interaction is that occurring between fast neutrons and 28Si, producing 28Al by the reaction 28Si(n,p)23Al. The product, 28Al, decays with a 2.3 minute half life, emitting 1.78 MeV gamma radiation. Similarly, when the aluminium constituent of ~he sample is irradiated with slow neutrons, the same radioactive isotope, 28Al, is produced with the consequent emission of l.78 MeV gamma radiation. Since there is negligible interaction between the fast neutrons and the material in small samples, the chemical concentrations of silicon and aluminium can be calibrated directly against the number of 1.78 MeV gamma ray counts observed within a given time interval after irradiation first with fast neutrons and then with slow neutrons. With bulk samples of bauxite, particularly samples having a significant water content, special allowance must be made for the slowing down of the fast neutrons used for the silicon analysis and the associated production of 28Al due to the capture of slow neutrons by the aluminium during the same irradiation.
One method of allowing for this effect is described below, in the description of the operation of the present in~ention.
For the currently used neutron activation analysis techniques, because two irradiations are necessary (with fast, then slow neutrons), there is a considerable 3~ capital outlay on the two neutron sources and their 115~377 respective shielding assemblies. If the analysis is applied to a moving stream of ore on a belt, two spectro-metric gamma ray detectors lfor example, 127 x 127 mm NaI(Tl)] are also required. If the analysis is performed on bulk samples contained in bins or boxes, although only one gamma ray de~tector is necessary, the analysis procedures for silicon and aluminium must duplicate each other, which doubles the necessary time and effort for analysis. In addition, for the measurements to be useful to the ana~yst, it is essential that the fast and slow neutron flux should be constant and reproducible from one measurement to the next.
The present invention offers appreciable savings in time and equipment cost, compared with current technology, by providing a method of analysis which is based on a single sample irradiation followed by a single measurement procedure.
The nuclear reactions providing the basis of the ~ present invention are:
There are two particular areas in the bauxite industry where the present invention will be used. One is the monitoring of ore quality during ship-loading operations of bauxite for export, where monitoring of aluminium grade and silicon impurity concentrations are essential to ensure that the ore satisfies export contract specifications.
The other use is the monitoring of ore quality whilst sorting the bauxite into stockpiles of different specified chemical concentrations of silicon and aluminium.
Depending on the way in which they have been formed, these stockpiles (a) may contain ore which has been blended , for ore treatment plants or (b~ may be used in subsequent blending operations.
In both of these situations, the current practice in monitoring the ore quality involves the periodic sampling of the ore from the bulk supply, which is usually moving on a conveyor belt when the sample is taken. The samples are moderately large (several kilogrammes) and are either subsampled immediately, or are mixed with other samples, taken by a predetermined number of automatic sampling cycles to form a representative bulk sample, which is then sub-sampled. Sub-sampling and crushing proceeds until a small specimen (of about 1 g) is ultimately available for chemical analysis by wet chemical assaying or by x-ray fluorescence analysis procedures. These sample preparation 115~377 procedures are particularly time-consuming if good representivity of the bulk is required in the sample.
The analysis is also time-consuming.
It has been found that in some situations (for instance during ship loading), when variations of ore quality occur these analytical methods are not fast enough to permit steps to be taken to correct the chemical concentrations of aluminium and silicon (for example, by further blending measures). If it were possible to apply the prior art techniques to on-stream analysis of bauxite on a moving belt, a more rapid analysis of the constituents and hence more rapid corrective blending measures should, in principle, be possible. Unfortunately, wet chemical methods cannot be applied to an on-stream situation, and on-stream x-ray fluorescence methods are inapplicable to lump-flow , measurement. The x-ray fluorescence method is also j unsuitable for the ahalysis of untreated bulk samples, due to the low penetration of x-radiation (less than 1 mm), and the fact that the ore is heterogeneous as regards moisture and particle size.
Neutron activation analysis, which is the basis of the present invention, does not have the problems associated with wet chemical assaying or the x-ray fluorescence technique, noted above, when applied to the analysis of large bulk samples. Indeed, activation analysis methods are directly applicable to large bulk samples and require minimal sample preparation in terms of crushing and drying. They also avoid most of the heterogeneity problems associated with the application of x-ray fluorescence analysis to bulk samples because the neutrons and gamma rays involved have a much deeper pehetration than x-rays. For this reason, the monitoring of bauxite ore ~uality (and the aluminium and silicon content of other materials) on moving belts is amenable to '` `' .:
'. ~ ':
- . .
115~377 the neutron activation method of the present invention.
Neutron activation methods have previously been applied to the analysis of silicon and aluminium in small samples (less than 150 g)-. For example, they ha~e been described in the paper by F. Dugain and J. Tatar in Ann. Inst. Geol. Publici Hungary, Volume 54, p 375 (i970), and in the papers ~y L. Alaerts, J.P. Op de Beeck, and J. Hoste, in Anal. Chim. Acta, Volume 70, p 253 (1974) and in Anal. Chim. Acta, Volume 78, p ~29 (1975). These methods depend on two interactions with the constituent chemical elements. One interaction is that occurring between fast neutrons and 28Si, producing 28Al by the reaction 28Si(n,p)23Al. The product, 28Al, decays with a 2.3 minute half life, emitting 1.78 MeV gamma radiation. Similarly, when the aluminium constituent of ~he sample is irradiated with slow neutrons, the same radioactive isotope, 28Al, is produced with the consequent emission of l.78 MeV gamma radiation. Since there is negligible interaction between the fast neutrons and the material in small samples, the chemical concentrations of silicon and aluminium can be calibrated directly against the number of 1.78 MeV gamma ray counts observed within a given time interval after irradiation first with fast neutrons and then with slow neutrons. With bulk samples of bauxite, particularly samples having a significant water content, special allowance must be made for the slowing down of the fast neutrons used for the silicon analysis and the associated production of 28Al due to the capture of slow neutrons by the aluminium during the same irradiation.
One method of allowing for this effect is described below, in the description of the operation of the present in~ention.
For the currently used neutron activation analysis techniques, because two irradiations are necessary (with fast, then slow neutrons), there is a considerable 3~ capital outlay on the two neutron sources and their 115~377 respective shielding assemblies. If the analysis is applied to a moving stream of ore on a belt, two spectro-metric gamma ray detectors lfor example, 127 x 127 mm NaI(Tl)] are also required. If the analysis is performed on bulk samples contained in bins or boxes, although only one gamma ray de~tector is necessary, the analysis procedures for silicon and aluminium must duplicate each other, which doubles the necessary time and effort for analysis. In addition, for the measurements to be useful to the ana~yst, it is essential that the fast and slow neutron flux should be constant and reproducible from one measurement to the next.
The present invention offers appreciable savings in time and equipment cost, compared with current technology, by providing a method of analysis which is based on a single sample irradiation followed by a single measurement procedure.
The nuclear reactions providing the basis of the ~ present invention are:
2~-~ 27Al(n,p)27Mg (used for the determination of the aluminium constituent), ~~-and 28Si(n,p)28A-l (for the silicon determination).
The energies of neutrons effective in these reactions are greater than 4.5 MeV. The radioactive nucleus 27Mg decays with a half life of 9.46 minutes and emits two gamma rays during its decay, which have energies of 0.844 MeV and 1.055 MeV respectively. The emission and half life of the other radioactive nucleus, 28Al, have been described above. As previously mentioned, a third nuclear reaction is important with all bulk samples having significant water content as well as aluminium and silicon constituents. This reaction, 27Al(n,y)28Al, which entails the capture of slow neutrons in aluminium, results in the emission of 1.78 MeV gamma radiation 1`156377 which is additional to the 1.78 MeV gamma radiation resulting from the fast neutron reaction with the silicon constituent of the sample. (Note that even with sources emitting only fàst neutrons for sample irradiations, appreciable numbers of fast neutrons are slowed down to thermal energies within the sample by their collision with the hydrogen nuclei associated with the water content of the sample).
Applications of the above nuclear reactions for the fast neutron activation analysis of aluminium and silicon have been described in the scientific literature~
For example, reference can be made to the paper by R.H.
Gijbels and J. Hertogen in Pure Appl. Chem., Volume 49, p 1555, (1977), and the paper by J. Kuusi in Nucl. Appl.
lS ~echnol., Volume-8, p 465 (1970). However, these applications are either for small samples, or for larger samples that contain little hydrogen and therefore cause negligible moderation of the fast neutrons within the samples.
The present invention overcomes the problem of interference by aluminium with silicon from the 1.78 MeV
gamma radiation in the following way~ After the fast neutron irradiation, the 1.78 MeV gamma rays from the sample are measured concurrently with those emitted by 27Mg at 0.844 MeV and 1.015 MeV for a preset time interval. (If the sample container is fabricated from A material~suc~-as copper, which produces gamma radiation-interfering with the 1.015 MeV gamma rays of the sample, the 1.015 MeV gamma rays are excluded from the analysis).
Since the number of counts from 27Mg are due only to aluminium, the chemical concentration of aluminium can be related directly to these recorded counts, given a knowledge of the mass of the sample. With a knowledge of the aluminium content of the sample, provided the slow neutron ~lux in the material is also known, the component ..... . .
of the 1.78 MeV gamma radiation due to slow neutron . reactions with the aluminium can be subtracted from the total 1.78 MeV gamma radiation count to provide the gamma radiation at 1.78 MeV resulting from fast neutron activation of the silicon in the sample. -Because the thermal neutron flux within the bul~
sample is sensitive to water content, it is essential to measure the number of thermal neutrons in a gi~en time interval during the neutron irradiation.
For this purpose, a sùitable neutron detector will be located adjacent to the sample. The number of neutrons recorded by the detector is proportional to the thermal neutron flux within the sample.
Thus, according to the present invention, a method - 15 of simultaneously analysing the aluminium and siliconcontent of a sample of material comprises the steps of:
(a) irradiating the sample with fast neutrons;
(b) monitoring the thermal neutron flux within the sample;
(c) monitoring the gamma radiation from the irradiated sample at enerqies of 1.78 MeV
and 1.015 and/or 0.844 MeV;
- ~d) using the monitored gamma radiation at 1.015 and/or 0.844 MeV to estimate the aluminium content of the sample; and (e) using the monitored gamma radiation at 1.78 MeV, .~ .. r ~ r ). ~1 ~compensated ~y the gamma radiation at l.78 MeV -~
due to the thermal neutron reaction with the estimated aluminium in the sample, to estimate the silicon content of the sample.
Also according to the present invention, apparatus for the simultaneous analysis of aluminium and silicon content of a sample of material comprises:
(a) a fast neutron source, adapted to irradiate the . .r 35 ~ s-ample of material; --~7~
l 15~377 (b) a thermal neutron detector, located to monitor the thermal neutron flux in the irradiated sample; and (c) a gamma ray detector separated from the neutron source and shielded therefrom, adapted to monitor the gamma spectrum from the irradiated sample, at least at energies of 1.78 MeV and of 0.844 and/or 1.015 MeV.
Other features of the present invention will become apparent from the following description of an embodiment of the invention, in which reference will be made to the accompanying drawings, of which:
Figure 1 is a schematic diagram of an experimental arrangement for the measurement of the aluminium and silicon content of a sample:
Figure 2 is a diagram, partly schematic and partly in block form, illustrating the components used in the neutron activation analysis arrangement illustrated in Figure l; and 1, Figures 3 and 4 are graphs displaying the results of aluminium and silicon determinations in samples which have been subjected to both conventional analysis and analysis by the neutron activation technique of the present invention.
Before describing the apparatus and technique in detail it will be helpful to consider the mathematics associated with the present invention.
' 30 In a sample containing both aluminium and silicon, j the grade of aluminium, Al, is related to the number of ;~ counts, G, recorded of gamma rays emitted by 27Mg at 0.844 MeV and/or 1.015 MeV, and to the sample weight, W, by the equation:
35 Al = a + alG + a2W. (1) . .
' ~
W
:~
' ' ' : -- .
' -` 1156377 The constant coefficients aO, al and a2 are determined from linear regression analysis by calibrating the responses G and W of the apparatus against the aluminium content of known samples using linear regression analysis.
The number of counts G is determined from the equation T ' (2) where GT is the total number of gamma rays recorded in - an energy window encompassing the 0.844 MeV and/or 1.015 peaks, J is the total number of counts recorded of 1.78 MeV gamma rays, and k is a constant. The term kJ is used to subtract the spectral continuum due to both the Compton scattered 1.78 MeV gamma radiation and background due to the neutron source.
Thus Al = aO + alGT + a2W + a3J, (3)
The energies of neutrons effective in these reactions are greater than 4.5 MeV. The radioactive nucleus 27Mg decays with a half life of 9.46 minutes and emits two gamma rays during its decay, which have energies of 0.844 MeV and 1.055 MeV respectively. The emission and half life of the other radioactive nucleus, 28Al, have been described above. As previously mentioned, a third nuclear reaction is important with all bulk samples having significant water content as well as aluminium and silicon constituents. This reaction, 27Al(n,y)28Al, which entails the capture of slow neutrons in aluminium, results in the emission of 1.78 MeV gamma radiation 1`156377 which is additional to the 1.78 MeV gamma radiation resulting from the fast neutron reaction with the silicon constituent of the sample. (Note that even with sources emitting only fàst neutrons for sample irradiations, appreciable numbers of fast neutrons are slowed down to thermal energies within the sample by their collision with the hydrogen nuclei associated with the water content of the sample).
Applications of the above nuclear reactions for the fast neutron activation analysis of aluminium and silicon have been described in the scientific literature~
For example, reference can be made to the paper by R.H.
Gijbels and J. Hertogen in Pure Appl. Chem., Volume 49, p 1555, (1977), and the paper by J. Kuusi in Nucl. Appl.
lS ~echnol., Volume-8, p 465 (1970). However, these applications are either for small samples, or for larger samples that contain little hydrogen and therefore cause negligible moderation of the fast neutrons within the samples.
The present invention overcomes the problem of interference by aluminium with silicon from the 1.78 MeV
gamma radiation in the following way~ After the fast neutron irradiation, the 1.78 MeV gamma rays from the sample are measured concurrently with those emitted by 27Mg at 0.844 MeV and 1.015 MeV for a preset time interval. (If the sample container is fabricated from A material~suc~-as copper, which produces gamma radiation-interfering with the 1.015 MeV gamma rays of the sample, the 1.015 MeV gamma rays are excluded from the analysis).
Since the number of counts from 27Mg are due only to aluminium, the chemical concentration of aluminium can be related directly to these recorded counts, given a knowledge of the mass of the sample. With a knowledge of the aluminium content of the sample, provided the slow neutron ~lux in the material is also known, the component ..... . .
of the 1.78 MeV gamma radiation due to slow neutron . reactions with the aluminium can be subtracted from the total 1.78 MeV gamma radiation count to provide the gamma radiation at 1.78 MeV resulting from fast neutron activation of the silicon in the sample. -Because the thermal neutron flux within the bul~
sample is sensitive to water content, it is essential to measure the number of thermal neutrons in a gi~en time interval during the neutron irradiation.
For this purpose, a sùitable neutron detector will be located adjacent to the sample. The number of neutrons recorded by the detector is proportional to the thermal neutron flux within the sample.
Thus, according to the present invention, a method - 15 of simultaneously analysing the aluminium and siliconcontent of a sample of material comprises the steps of:
(a) irradiating the sample with fast neutrons;
(b) monitoring the thermal neutron flux within the sample;
(c) monitoring the gamma radiation from the irradiated sample at enerqies of 1.78 MeV
and 1.015 and/or 0.844 MeV;
- ~d) using the monitored gamma radiation at 1.015 and/or 0.844 MeV to estimate the aluminium content of the sample; and (e) using the monitored gamma radiation at 1.78 MeV, .~ .. r ~ r ). ~1 ~compensated ~y the gamma radiation at l.78 MeV -~
due to the thermal neutron reaction with the estimated aluminium in the sample, to estimate the silicon content of the sample.
Also according to the present invention, apparatus for the simultaneous analysis of aluminium and silicon content of a sample of material comprises:
(a) a fast neutron source, adapted to irradiate the . .r 35 ~ s-ample of material; --~7~
l 15~377 (b) a thermal neutron detector, located to monitor the thermal neutron flux in the irradiated sample; and (c) a gamma ray detector separated from the neutron source and shielded therefrom, adapted to monitor the gamma spectrum from the irradiated sample, at least at energies of 1.78 MeV and of 0.844 and/or 1.015 MeV.
Other features of the present invention will become apparent from the following description of an embodiment of the invention, in which reference will be made to the accompanying drawings, of which:
Figure 1 is a schematic diagram of an experimental arrangement for the measurement of the aluminium and silicon content of a sample:
Figure 2 is a diagram, partly schematic and partly in block form, illustrating the components used in the neutron activation analysis arrangement illustrated in Figure l; and 1, Figures 3 and 4 are graphs displaying the results of aluminium and silicon determinations in samples which have been subjected to both conventional analysis and analysis by the neutron activation technique of the present invention.
Before describing the apparatus and technique in detail it will be helpful to consider the mathematics associated with the present invention.
' 30 In a sample containing both aluminium and silicon, j the grade of aluminium, Al, is related to the number of ;~ counts, G, recorded of gamma rays emitted by 27Mg at 0.844 MeV and/or 1.015 MeV, and to the sample weight, W, by the equation:
35 Al = a + alG + a2W. (1) . .
' ~
W
:~
' ' ' : -- .
' -` 1156377 The constant coefficients aO, al and a2 are determined from linear regression analysis by calibrating the responses G and W of the apparatus against the aluminium content of known samples using linear regression analysis.
The number of counts G is determined from the equation T ' (2) where GT is the total number of gamma rays recorded in - an energy window encompassing the 0.844 MeV and/or 1.015 peaks, J is the total number of counts recorded of 1.78 MeV gamma rays, and k is a constant. The term kJ is used to subtract the spectral continuum due to both the Compton scattered 1.78 MeV gamma radiation and background due to the neutron source.
Thus Al = aO + alGT + a2W + a3J, (3)
3 lK.
Similarly, the chemical concentration of silicon, Si, in the sample can be related to the number of counts per unit time, H, of 1.78 MeV gamma rays due to the 28Si(n,p)28Al reaction and the sample weight, W, ~y:
Si = bo + bl~ + b2W, (4) where bo, bl and b2 are constant coefficients obtainable from regression analysis ~y calibrating the responses H
and W against the silicon content of known samples using linear regression analysis.
In practice, with bulk bauxite samples, however, the total number of counts per unit time due to 1.78 MeV gamma rays from 28Al, J, contains two indistinguisha~le components H and I, where I is the component contributed by thermal neutron actiYation of 27Al. The number of counts, I, is proportional to the product of the number of thermal neutrons, Nt, measured during irradiation and the aluminium concentration of the sample. Because the number of counts, GT, pre~iously referred to, are related to the aluminium _ g _ B
l 15~377 concentratlon, and since Nt is proportional to the thermal neutron flux within the sample, the actual silicon concentration in the sample is given by: :
bo + b3J + b4GTNt + b2W' (5) where b3 and b4 are regression coefficients, and where bo~ b2, b3 and b4 are calibration coefficients determined by linear regression analysis as described above.
- ga -~B
, .
115~377 Equations (3) and (5j are used in analysis of materials in accordance with the present invention.
The experimental arrangement devised to test the present invention, and illustrated in Figure 1, comprises a neutron source 10 and thermal neutron detector 11 located close to a railway track 12 on which a small sample of material 13 can be moved. Also close to the track 12, but remote from the neutron source 10 and detector 11, is a gamma ray detector 14, suitably shielded by a lead screen 15 and a masonry shield 16 .
In the experimental facility shown in Figure 1, the neutron source was 20 Ci of Am-Be, giving an estimated output of 4.4 x 107 n/s. The samples of material 13 were contained in a rectangular brass box 17 (25 x 25 x 4 cm deep) and were irradiated by fast neutrons from underneath. The neutron source 10 was enclosed within a cylindrical shell 21 of cadmium (see Figure 2) which prevented the thermal neutrons emitted by the source from reaching the sample 13. The thermal neutron flux generated within the sample 13 was monitored by thermal neutron detector 11, which was a high efficiency neutron detector (filled to a pressure of 4 atmospheres with a mixture of 3He and Krj which was also located beneath the sample container and adjacent to the neutron source.
After irradiation by source 10 for 6 minutes, the sample 13 was transferred within 15 seconds to a position immediately above the gamma detector 14, which comprised a 127 x 127 mm NaI(Tl) scintillation detector. The distance between the neutron source 10 and gamma detector 14 was about 7 metres. This separation between source and detector added a considerable comp~nent of distance shielding to the already appreciable shielding against source radiation provided by concrete bric~
structures 16 and 26 built around both the NaI(Tl) detector 14 and the neutron source assemblies. The 115~377 lead shield 15, of thickness 3cm, which was built around the body of the scintillation detector 14 so as -to leave only the-upper plane surface exposed for measurements, provided further reduction of background radiation. Spectrum stabilization was obtained using O.662 MeV gamma rays from a 137Cs source (not shown in the drawings) which provided a reference peak for a Canberra Industries Model 1520 analogue spectrum stabilizer.
The energy spectra of gamma rays detected by the scintillation detector 14 were analysed in the initial phases of the development of the method by a Tracor Northern 4096-channel pulse height analyser (model TN-1700). At a later stage, when energy-pulse height calibrations had been fully established, Ortec single ~5 channel analysers, digital counters and a timer were used, as shown in Figure 2, for their greater suitability to plant or mine site operating requirements. The amplifiers used were a Tennelec linear amplifier for the scintillation detector 14 and an Ortec spectroscopy amplifier for the neutron detector 11. Output from the digital counters was obtained with a strip printer.
! The partly schematic and partly block form diagram of Figure 2 is essentially a more comprehensive illustration of the apparatus shown in Figure 1. In particular, the high energy neutron source 10 and the thermal neutron detector 11 are shown more explicitly, with the neutron source 10 encased in ca& ium shielding 21 and the source 10 and detector 11 positioned within masonry shielding 26.
A single high voltage power supply 22 services both the thermal neutron detector 11 and the gamma ray detector 14.
The output signal from the thermal neutron detector 11 is first amp~ified in gain control unit 23, and if the signal, when amplified, exceeds a required threshold value (determined by threshold device 24), it is supplied to the input of a digital counter ~5. The output from counter 25 is fed into processor 32, which is usually a microprocessor or small computer, programmed to effect the required analysis from its three input signals.
The other two input signals to the processor 32 are signals indicative of the values G and J (see the above description). These signals are derived from the gamma rays received by the gamma detector 14 as a result of the decay of, respectively, the 23Al and 27Mg isotopes formed during the irradiation of the sample. These inputs are obtained after the output of detector 14 has been processed by amplifier 26, a gamma ray discriminator, and digital counters 28 and 28A. The gamma ray discriminator has been shown in the drawing as two gamma ray single channel analysers 27 and 27A. In practice, these devices 27 and 27A may be a single unit comprising a multi-channel analyser with outputs from channels which have energy windows, typically about 0.35 MeV wide, centred on 0.844 MeV, 1.015 MeV and 1.78 MeV.
Presettable timer 31 controls the operation of the digital counters 28 and 28A. Timer 31 will be synchronized with, but operating sequentially to, pre-settable timer 30 which controls the operation of the digital counter 25.
The output from the processor 32 may be recorded.
For example, it may be stored on magnetic tape, magnetic disc, magnetic card, punched tape, punched card, or on any other suitable,medium. Alternatively, or additionally, the output from the processor 32 may be presented as a digital display, a paper print-out, or on a chart recorder. Those skilled in this fieId will appreciate that the actual form of the presentation of the output from processor 32 may be chosen to suit the requirements of the owner or operator of the equipment~ Accordingly, a single, unspecified display unit 33 has been included in Figure 2.
-~ ` 35 --- In one example of the experimental testing of the 1 15~377 present invention, bauxite samples were dried to less than 5 per cent (by weight) free moisture and crushed to -6 mm particle size. It should be noted, however, that this amount of pre-treatment is not essential.
- 5 The bauxite samples contained aluminium in the range from 26 wt. per cent ~o 32 wt. per cent, whilst the silicon concentrations ranged from 0.9 wt. per cent to
Similarly, the chemical concentration of silicon, Si, in the sample can be related to the number of counts per unit time, H, of 1.78 MeV gamma rays due to the 28Si(n,p)28Al reaction and the sample weight, W, ~y:
Si = bo + bl~ + b2W, (4) where bo, bl and b2 are constant coefficients obtainable from regression analysis ~y calibrating the responses H
and W against the silicon content of known samples using linear regression analysis.
In practice, with bulk bauxite samples, however, the total number of counts per unit time due to 1.78 MeV gamma rays from 28Al, J, contains two indistinguisha~le components H and I, where I is the component contributed by thermal neutron actiYation of 27Al. The number of counts, I, is proportional to the product of the number of thermal neutrons, Nt, measured during irradiation and the aluminium concentration of the sample. Because the number of counts, GT, pre~iously referred to, are related to the aluminium _ g _ B
l 15~377 concentratlon, and since Nt is proportional to the thermal neutron flux within the sample, the actual silicon concentration in the sample is given by: :
bo + b3J + b4GTNt + b2W' (5) where b3 and b4 are regression coefficients, and where bo~ b2, b3 and b4 are calibration coefficients determined by linear regression analysis as described above.
- ga -~B
, .
115~377 Equations (3) and (5j are used in analysis of materials in accordance with the present invention.
The experimental arrangement devised to test the present invention, and illustrated in Figure 1, comprises a neutron source 10 and thermal neutron detector 11 located close to a railway track 12 on which a small sample of material 13 can be moved. Also close to the track 12, but remote from the neutron source 10 and detector 11, is a gamma ray detector 14, suitably shielded by a lead screen 15 and a masonry shield 16 .
In the experimental facility shown in Figure 1, the neutron source was 20 Ci of Am-Be, giving an estimated output of 4.4 x 107 n/s. The samples of material 13 were contained in a rectangular brass box 17 (25 x 25 x 4 cm deep) and were irradiated by fast neutrons from underneath. The neutron source 10 was enclosed within a cylindrical shell 21 of cadmium (see Figure 2) which prevented the thermal neutrons emitted by the source from reaching the sample 13. The thermal neutron flux generated within the sample 13 was monitored by thermal neutron detector 11, which was a high efficiency neutron detector (filled to a pressure of 4 atmospheres with a mixture of 3He and Krj which was also located beneath the sample container and adjacent to the neutron source.
After irradiation by source 10 for 6 minutes, the sample 13 was transferred within 15 seconds to a position immediately above the gamma detector 14, which comprised a 127 x 127 mm NaI(Tl) scintillation detector. The distance between the neutron source 10 and gamma detector 14 was about 7 metres. This separation between source and detector added a considerable comp~nent of distance shielding to the already appreciable shielding against source radiation provided by concrete bric~
structures 16 and 26 built around both the NaI(Tl) detector 14 and the neutron source assemblies. The 115~377 lead shield 15, of thickness 3cm, which was built around the body of the scintillation detector 14 so as -to leave only the-upper plane surface exposed for measurements, provided further reduction of background radiation. Spectrum stabilization was obtained using O.662 MeV gamma rays from a 137Cs source (not shown in the drawings) which provided a reference peak for a Canberra Industries Model 1520 analogue spectrum stabilizer.
The energy spectra of gamma rays detected by the scintillation detector 14 were analysed in the initial phases of the development of the method by a Tracor Northern 4096-channel pulse height analyser (model TN-1700). At a later stage, when energy-pulse height calibrations had been fully established, Ortec single ~5 channel analysers, digital counters and a timer were used, as shown in Figure 2, for their greater suitability to plant or mine site operating requirements. The amplifiers used were a Tennelec linear amplifier for the scintillation detector 14 and an Ortec spectroscopy amplifier for the neutron detector 11. Output from the digital counters was obtained with a strip printer.
! The partly schematic and partly block form diagram of Figure 2 is essentially a more comprehensive illustration of the apparatus shown in Figure 1. In particular, the high energy neutron source 10 and the thermal neutron detector 11 are shown more explicitly, with the neutron source 10 encased in ca& ium shielding 21 and the source 10 and detector 11 positioned within masonry shielding 26.
A single high voltage power supply 22 services both the thermal neutron detector 11 and the gamma ray detector 14.
The output signal from the thermal neutron detector 11 is first amp~ified in gain control unit 23, and if the signal, when amplified, exceeds a required threshold value (determined by threshold device 24), it is supplied to the input of a digital counter ~5. The output from counter 25 is fed into processor 32, which is usually a microprocessor or small computer, programmed to effect the required analysis from its three input signals.
The other two input signals to the processor 32 are signals indicative of the values G and J (see the above description). These signals are derived from the gamma rays received by the gamma detector 14 as a result of the decay of, respectively, the 23Al and 27Mg isotopes formed during the irradiation of the sample. These inputs are obtained after the output of detector 14 has been processed by amplifier 26, a gamma ray discriminator, and digital counters 28 and 28A. The gamma ray discriminator has been shown in the drawing as two gamma ray single channel analysers 27 and 27A. In practice, these devices 27 and 27A may be a single unit comprising a multi-channel analyser with outputs from channels which have energy windows, typically about 0.35 MeV wide, centred on 0.844 MeV, 1.015 MeV and 1.78 MeV.
Presettable timer 31 controls the operation of the digital counters 28 and 28A. Timer 31 will be synchronized with, but operating sequentially to, pre-settable timer 30 which controls the operation of the digital counter 25.
The output from the processor 32 may be recorded.
For example, it may be stored on magnetic tape, magnetic disc, magnetic card, punched tape, punched card, or on any other suitable,medium. Alternatively, or additionally, the output from the processor 32 may be presented as a digital display, a paper print-out, or on a chart recorder. Those skilled in this fieId will appreciate that the actual form of the presentation of the output from processor 32 may be chosen to suit the requirements of the owner or operator of the equipment~ Accordingly, a single, unspecified display unit 33 has been included in Figure 2.
-~ ` 35 --- In one example of the experimental testing of the 1 15~377 present invention, bauxite samples were dried to less than 5 per cent (by weight) free moisture and crushed to -6 mm particle size. It should be noted, however, that this amount of pre-treatment is not essential.
- 5 The bauxite samples contained aluminium in the range from 26 wt. per cent ~o 32 wt. per cent, whilst the silicon concentrations ranged from 0.9 wt. per cent to
4.5 wt. per cent. The mass of the sample used for irradiation was about 4 kg.
As expected, when the bauxite was irradiated with fast neutrons, the gamma ray spectra were dominated by the 1.78 MeV gamma ray peak due to 28Al, and by a spectral continuum of gamma rays which had undergone Compton scattering both within the detector and within the bulk sample. This continuum underlies the spectral peaks at 1.015 and 0.844 MeV due to 27Mg. The Compton scattering processes in this example were dominated by gamma rays which initially had energies of 1.78 MeV, 1.014 MeV and 0.844 MeV originating from the sample, and 0.662 Me~ due to the 137Cs stabilization reference source. In the case of bauxite, again as expected, the interferences from other constituents, such as the natural radioactive nuclides, was minimal, and those from 56Mn at 0.846 MeU, 1.811 MeV and 2.113 MeV were very small.
(This nuclide, with a 2.57 hr. half life, can arise from the 55Mn(n,r)56Mn reaction, or from a 56Fe(n,p)56Mn reaction; the first of these réactions will contribute negligible 56Mn owing to the extremely low concentration of manganese in Australian bauxites, despite the relatively large cross section of 13.3 barns for that reaction; the second rea~;on, involving iron, contributes more 56Mn than the first, but constitutes a constant level of about 2 per cent interference, the vari~tion of which - is only about 1 per cent of the gamma ray signal from 27Mg.) 35~ An~her~source of spectral inteEference-which occurred in this example arose from the neutron activation .. . .
115~377 of the copper constituent of the brass sample container, which`contributed a small peak at 1.05 MeV. -`
- It was necessa~y, the~ef-ore,- to exclude from-calcula'tions-all count data that would hàve been recorded in a narrow energy win'dow, about 0.1 MeV wide, centred at 1.05 MeV.
Apart from inter~erences to the spectral peaks due to 28Al and:27Mg from monoenergetic gamma rays emitted by minor constituents of the sample and sample container, there was also substantial interference from the continuum of scatterèd gamma radiation. The extent of - --- - inter~erence-with-*he~ .78 MeV spectral peak appeared to '~ be insignificant owing to-negligible gamma radiation , apparent at higher energies. However, the 0.84i MeV
' gamma ray peak due to 27Mg received considerable - i-nter~erence f~om-the substantial underlying continuum' caused by Compton scattering of the 1.78 MeV gamma radiation from the decay of 23Al and background from the neutron source.
One technique that could have been used to overcome the interference problem when using multichannel pulse height analysers for neutron activation analysis is that ' - which is described in the specification of Australian~ -- patent No. 468,970. That method entails an estimation'' of the underlying continuum which is based on the number of counts in an energy channel close to the relevant ~ ) , . .
spectral peak. However, in the present experimental arrangement,~an alternative-method was effectively ~'~'~~'`'''~' , implemented with the use of single channel anal~sers for the activation analysis of bauxite. The method simply ~ 30 entailed the establishment of two particular energy i'¦ windows~ One window, centred at 0.844 MeV, is '~ approximately 0.1 MeV wide. The other window, about - ' 0.35 MeV wide, encompasses the 1.73 MeV peak.
Implementation of these two energy-window conditions ~i ~~ 3~ alone worked well because the counts-accumulatéd within .. . .
, -14-i!
!
. ' ' '~, ' ' , ' ' ll5~377 the spectral continuum occurring within the first narrow window are proportional, with good approximation, to the number of counts due to 28Al, 1.78 MeV gamma radiation. The counts recorded in these two windows were respectively denoted by GT and J in equations (3) and (5) for purposes of either determining the calibration coefficients, ai and bj, or for determining the chemical concentrations of silicon and aluminium in samples when calibrations, and hence coefficients, were already known.
After performing a number of experiments with well-blended, effectively homogeneous, ore samples of accurately known compositio~ the data from the activation analysis were fitted against the ~nown chemical assays for aluminium and silicon by linear regression analysis in order to determine the constant coefficients in equations (3) and
As expected, when the bauxite was irradiated with fast neutrons, the gamma ray spectra were dominated by the 1.78 MeV gamma ray peak due to 28Al, and by a spectral continuum of gamma rays which had undergone Compton scattering both within the detector and within the bulk sample. This continuum underlies the spectral peaks at 1.015 and 0.844 MeV due to 27Mg. The Compton scattering processes in this example were dominated by gamma rays which initially had energies of 1.78 MeV, 1.014 MeV and 0.844 MeV originating from the sample, and 0.662 Me~ due to the 137Cs stabilization reference source. In the case of bauxite, again as expected, the interferences from other constituents, such as the natural radioactive nuclides, was minimal, and those from 56Mn at 0.846 MeU, 1.811 MeV and 2.113 MeV were very small.
(This nuclide, with a 2.57 hr. half life, can arise from the 55Mn(n,r)56Mn reaction, or from a 56Fe(n,p)56Mn reaction; the first of these réactions will contribute negligible 56Mn owing to the extremely low concentration of manganese in Australian bauxites, despite the relatively large cross section of 13.3 barns for that reaction; the second rea~;on, involving iron, contributes more 56Mn than the first, but constitutes a constant level of about 2 per cent interference, the vari~tion of which - is only about 1 per cent of the gamma ray signal from 27Mg.) 35~ An~her~source of spectral inteEference-which occurred in this example arose from the neutron activation .. . .
115~377 of the copper constituent of the brass sample container, which`contributed a small peak at 1.05 MeV. -`
- It was necessa~y, the~ef-ore,- to exclude from-calcula'tions-all count data that would hàve been recorded in a narrow energy win'dow, about 0.1 MeV wide, centred at 1.05 MeV.
Apart from inter~erences to the spectral peaks due to 28Al and:27Mg from monoenergetic gamma rays emitted by minor constituents of the sample and sample container, there was also substantial interference from the continuum of scatterèd gamma radiation. The extent of - --- - inter~erence-with-*he~ .78 MeV spectral peak appeared to '~ be insignificant owing to-negligible gamma radiation , apparent at higher energies. However, the 0.84i MeV
' gamma ray peak due to 27Mg received considerable - i-nter~erence f~om-the substantial underlying continuum' caused by Compton scattering of the 1.78 MeV gamma radiation from the decay of 23Al and background from the neutron source.
One technique that could have been used to overcome the interference problem when using multichannel pulse height analysers for neutron activation analysis is that ' - which is described in the specification of Australian~ -- patent No. 468,970. That method entails an estimation'' of the underlying continuum which is based on the number of counts in an energy channel close to the relevant ~ ) , . .
spectral peak. However, in the present experimental arrangement,~an alternative-method was effectively ~'~'~~'`'''~' , implemented with the use of single channel anal~sers for the activation analysis of bauxite. The method simply ~ 30 entailed the establishment of two particular energy i'¦ windows~ One window, centred at 0.844 MeV, is '~ approximately 0.1 MeV wide. The other window, about - ' 0.35 MeV wide, encompasses the 1.73 MeV peak.
Implementation of these two energy-window conditions ~i ~~ 3~ alone worked well because the counts-accumulatéd within .. . .
, -14-i!
!
. ' ' '~, ' ' , ' ' ll5~377 the spectral continuum occurring within the first narrow window are proportional, with good approximation, to the number of counts due to 28Al, 1.78 MeV gamma radiation. The counts recorded in these two windows were respectively denoted by GT and J in equations (3) and (5) for purposes of either determining the calibration coefficients, ai and bj, or for determining the chemical concentrations of silicon and aluminium in samples when calibrations, and hence coefficients, were already known.
After performing a number of experiments with well-blended, effectively homogeneous, ore samples of accurately known compositio~ the data from the activation analysis were fitted against the ~nown chemical assays for aluminium and silicon by linear regression analysis in order to determine the constant coefficients in equations (3) and
(5). The respective precisions for silicon and aluminium determinations in bulk samples were obtained in terms of the sample standard deviations (s~ as shown below:
(al When using equations (3~ and (5), and the method of the present invention, for Al: s = 0.43 per cent Al for Si: s = 0.14 per cent Si (b~ When the contribution by gamma rays from 27Mg at 0.844 MeV is omitted from equation 5, for Si: s = 0.19 per cent Si (c) When the contributions both by the gamma rays from 7Mg at 0.844 MeV and thermal neutrons measured below the sample container are omitted from equation 5, for Si: s = 0.82 per cent Si As shown by the smaller standard deviations for the results obtained using the present invention, the present invention compares most favourably with alternatives tb~ and tCI.
115~377 Comparisons between neutron activation determinations for aluminium and silicon, expressed as alumina (A12O~) and silica (SiO2) respectively, and determinations by conventional analysis are shown in Figures 3 and 4. The calibration equations used to calculate the neutron activation determinations of alumina and silica in Figures 3 and 4 were as follows:
A1203 = 71.04 - 0.946 GT ~ 9.636W - 0.242J (6) SiO2 = 12.93 + 0.665J - 0.0477 GTNt ~ 2.61W (7) where J and GT are expressed in thousands of counts, Nt in millions of counts, and W in kilograms.
It will be clear to those skilled in ~his art that (,a) the container 17 need not be of brass and thus need not generate a significant component of the gamma spectra being studied, (b) the rail and bulk sample of the experimental arrangement described above can be substituted by a conveyor belt carrying ore ~or other material) between a neutron irradiation station and a downstream gamma monitoring station, to ena~le on-stream analysis for silicon and aluminiu,m of the material being carried by the belt, and (c) the rail and bulk sample of the experimental arrangement described above can be substituted by the stationary walls and surrounding rock of a borehole, and both the source and detector can be simultaneously moved in the borehole to enable borehole logging for silicon and aluminium. For such an arrangement, the high energy neutron source, the thermal neutron detector and the gamma ray detector will be mounted on a borehole probe, which can then be lowered into a borehole to any required position to analyse the rock surrounding the borehole. Normally the signal processîng equi,pment will not be included on the probe, but will be connected to the source and detectors by long cables.
,~ - 16 -
(al When using equations (3~ and (5), and the method of the present invention, for Al: s = 0.43 per cent Al for Si: s = 0.14 per cent Si (b~ When the contribution by gamma rays from 27Mg at 0.844 MeV is omitted from equation 5, for Si: s = 0.19 per cent Si (c) When the contributions both by the gamma rays from 7Mg at 0.844 MeV and thermal neutrons measured below the sample container are omitted from equation 5, for Si: s = 0.82 per cent Si As shown by the smaller standard deviations for the results obtained using the present invention, the present invention compares most favourably with alternatives tb~ and tCI.
115~377 Comparisons between neutron activation determinations for aluminium and silicon, expressed as alumina (A12O~) and silica (SiO2) respectively, and determinations by conventional analysis are shown in Figures 3 and 4. The calibration equations used to calculate the neutron activation determinations of alumina and silica in Figures 3 and 4 were as follows:
A1203 = 71.04 - 0.946 GT ~ 9.636W - 0.242J (6) SiO2 = 12.93 + 0.665J - 0.0477 GTNt ~ 2.61W (7) where J and GT are expressed in thousands of counts, Nt in millions of counts, and W in kilograms.
It will be clear to those skilled in ~his art that (,a) the container 17 need not be of brass and thus need not generate a significant component of the gamma spectra being studied, (b) the rail and bulk sample of the experimental arrangement described above can be substituted by a conveyor belt carrying ore ~or other material) between a neutron irradiation station and a downstream gamma monitoring station, to ena~le on-stream analysis for silicon and aluminiu,m of the material being carried by the belt, and (c) the rail and bulk sample of the experimental arrangement described above can be substituted by the stationary walls and surrounding rock of a borehole, and both the source and detector can be simultaneously moved in the borehole to enable borehole logging for silicon and aluminium. For such an arrangement, the high energy neutron source, the thermal neutron detector and the gamma ray detector will be mounted on a borehole probe, which can then be lowered into a borehole to any required position to analyse the rock surrounding the borehole. Normally the signal processîng equi,pment will not be included on the probe, but will be connected to the source and detectors by long cables.
,~ - 16 -
Claims (20)
PROPERTY OR PRIVILEGE IS CLAIMED ARE DEFINED AS FOLLOWS:
1. A method of simultaneously analysing the aluminium and silicon content of a sample of material using only a single irradiation from a single neutron source comprises the steps of:
(a) irradiating the sample only with fast neutrons from said single neutron source;
(b) monitoring the thermal neutron flux within the sample, which flux results from moderation of the fast neutrons within the sample material;
(c) monitoring the gamma radiation from the irradiated sample at 1.78 MeV and at an energy selected from the group consisting of (1) 1.015 MeV, (2) 0.844 MeV and (3) 1.015 and 0.844 MeV;
(d) using the monitored gamma radiation of energy selected from said group to estimate the aluminium content of the sample; and (e) using the monitored gamma radiation at 1.78 MeV, compensated by the gamma radiation at 1.78 MeV due to the thermal neutron reaction with the estimated aluminium in the sample, to estimate the silicon content of the sample.
(a) irradiating the sample only with fast neutrons from said single neutron source;
(b) monitoring the thermal neutron flux within the sample, which flux results from moderation of the fast neutrons within the sample material;
(c) monitoring the gamma radiation from the irradiated sample at 1.78 MeV and at an energy selected from the group consisting of (1) 1.015 MeV, (2) 0.844 MeV and (3) 1.015 and 0.844 MeV;
(d) using the monitored gamma radiation of energy selected from said group to estimate the aluminium content of the sample; and (e) using the monitored gamma radiation at 1.78 MeV, compensated by the gamma radiation at 1.78 MeV due to the thermal neutron reaction with the estimated aluminium in the sample, to estimate the silicon content of the sample.
2. A method as defined in claim 1, in which step (d) includes compensating the monitored gamma radiation at an energy selected from said group for Compton scattered 1.78 MeV gamma radiation and background gamma radiation.
3. A method as defined in claim 1 or claim 2, in which the determination of the aluminium content of the sample in step (d) is effected using the equation Al = a0 + a1GT + a2W + a3J
where a0, a1, a2 and a3 are calibration coefficients that are determined by calibrating the responses GT, W and J against the aluminium content of known samples using regression analysis, GT is the total number of gamma detector counts in unit time of gamma rays having an energy selected from said group, J is the total number of gamma detector counts in unit time of gamma rays having an energy of 1.78 MeV, and W
is the weight of the sample.
where a0, a1, a2 and a3 are calibration coefficients that are determined by calibrating the responses GT, W and J against the aluminium content of known samples using regression analysis, GT is the total number of gamma detector counts in unit time of gamma rays having an energy selected from said group, J is the total number of gamma detector counts in unit time of gamma rays having an energy of 1.78 MeV, and W
is the weight of the sample.
4. A method as defined in claim 1 or claim 2, in which the determination of the silicon content in step (e) is effected using the equation Si = b0 + b3J + b4GTNt + b2W, where b0, b2, b3 and b4 are calibration coefficients determined by calibrating the responses J, GTNt and W
against the silicon content of known samples using regression analysis, J is the total number of gamma detector counts in unit time of gamma rays having an energy of 1.78 MeV, GT is the total number of counts in unit time of gamma rays having an energy in the window selected from said group and Nt is the observed thermal neutron flux in the sample during its irradiation with fast neutrons.
against the silicon content of known samples using regression analysis, J is the total number of gamma detector counts in unit time of gamma rays having an energy of 1.78 MeV, GT is the total number of counts in unit time of gamma rays having an energy in the window selected from said group and Nt is the observed thermal neutron flux in the sample during its irradiation with fast neutrons.
5. A method as defined in claim 1, in which the sample is a bulk sample and in which steps (a) and (b) are carried out at a first location, steps (c) and (d) are carried out at the second location which is remote from said first location, and said bulk sample is rapidly moved from said first location to said second location when steps (a) and (b) have been completed.
6. A method as defined in claim 1, in which said sample is being carried on a moving conveyor belt.
7. A method as defined in claim 6, in which steps (a) and (b) are carried out at a first location above or below said conveyor and steps (c) and (d) are carried out at a second location above or below said conveyor, said second location being down- stream of said first location.
8. A method as defined in claim 1, in which said sample is the material in the wall of a borehole.
9. Apparatus for the simultaneous analysis of aluminium and silicon content of a sample of material using only a single irradiation from a single neutron source comprises:
(a) a fast neutron source means for irradiating the sample of material;
(b) a thermal neutron detector means located to monitor the thermal neutron flux in the irradiated sample which flux results from the moderation of the fast neutrons within the sample material; and (c) a gamma ray detector means separated from the neutron source and shielded therefrom, for monitoring the gamma spectrum from the irradiated sample at least at 1.78 MeV and at an energy selected from the group consisting of (1) 1.015 MeV, (2) 0.844 MeV and (3) 1.015 MeV
and 0.844 MeV.
(a) a fast neutron source means for irradiating the sample of material;
(b) a thermal neutron detector means located to monitor the thermal neutron flux in the irradiated sample which flux results from the moderation of the fast neutrons within the sample material; and (c) a gamma ray detector means separated from the neutron source and shielded therefrom, for monitoring the gamma spectrum from the irradiated sample at least at 1.78 MeV and at an energy selected from the group consisting of (1) 1.015 MeV, (2) 0.844 MeV and (3) 1.015 MeV
and 0.844 MeV.
10. Apparatus as defined in claim 9, including a sample container movable quickly from a first location, in which it is positioned in close prox-imity to said fast neutron source means and said thermal neutron detector means to a second location, remote from said first location, in which it is positioned in close proximity to said gamma ray detector means.
11. Apparatus as defined in claim 10, in which said container is mounted on a railway track.
12. Apparatus as defined in claim 10 or claim 11, wherein said sample container is a brass box.
13. Apparatus as defined in claim 9, in which said fast neutron source means and said thermal neutron detector means are mounted at a first location in close proximity to a conveyor belt adapted to transport said material, and said gamma ray detector means is mounted in close proximity to said conveyor belt in a second location which is down-stream of said first location.
14. Apparatus as defined in claim 9, in which the fast neutron source, the thermal neutron detector and the gamma ray detector are mounted on a borehole probe.
15. Apparatus as defined in claim 14, in which said fast neutron source is encased in a cadmium shell.
16. Apparatus as defined in claim 15 including spectrum stabiliser means for stabilizing the spectrum from the irradiated sample.
17. Apparatus as defined in claim 16 including plural channel analyser means for monitoring the output of said gamma ray detector means and having channels responsive to the output of said gamma ray detector means when gamma radiation within a plurality of pre-selected energy windows is incident thereon, said energy windows including energy windows encompassing 0.844 MeV, 1.015 MeV and 1.78 MeV.
18. Apparatus as defined in claim 17, including (a) a plurality of counters, each counter being associated with a respective output of one of the channels of said plural channel analyser means, (b), a display device, and (c)a microprocessor programmed to derive, from the output of each said counter, values of the aluminium and silicon content of a material and to cause said display device to display and/or record said values.
19. Apparatus as defined in claims 17 or 18 wherein said multiple channel analyser means comprises a plurality of single channel analysers.
20. Apparatus as defined in claims 17 or 18 wherein said multiple channel analyser means comprises a multi-channel analyser.
Applications Claiming Priority (2)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| AUPD9897/79 | 1979-08-06 | ||
| AUPD989779 | 1979-08-06 |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CA1156377A true CA1156377A (en) | 1983-11-01 |
Family
ID=3768197
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CA000357743A Expired CA1156377A (en) | 1979-08-06 | 1980-08-06 | Neutron activation analysis |
Country Status (6)
| Country | Link |
|---|---|
| US (1) | US4361534A (en) |
| EP (1) | EP0024835B1 (en) |
| CA (1) | CA1156377A (en) |
| DE (1) | DE3068109D1 (en) |
| FI (1) | FI73527C (en) |
| ZA (1) | ZA804773B (en) |
Families Citing this family (20)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| HU181357B (en) * | 1981-06-25 | 1983-07-28 | Magyar Allami Eoetvoes Lorand | Geophysical nuclear analyzer, in particular to the analysis of mineral materials |
| US4464330A (en) * | 1982-05-13 | 1984-08-07 | The United States Of America As Represented By The Department Of Energy | Apparatus for irradiating a continuously flowing stream of fluid |
| GB8523060D0 (en) * | 1985-09-18 | 1985-10-23 | Cogent Ltd | Coal analysis |
| US5006299A (en) * | 1987-05-26 | 1991-04-09 | Science Applications International Corporation | Explosive detection system |
| US5162095A (en) * | 1987-07-15 | 1992-11-10 | L'etat Francais | Method and installation for the analysis by neutron activation of a flow of material in bulk |
| US5825030A (en) * | 1997-03-20 | 1998-10-20 | Gamma-Metrics | Shaping neutron energies to achieve sensitivity and uniformity of bulk material analysis |
| US20030165212A1 (en) * | 1998-02-18 | 2003-09-04 | Maglich Bogdan C. | Method and apparatus for detecting, locating, and analyzing chemical compounds using subatomic particle activation |
| US6438189B1 (en) * | 1998-07-09 | 2002-08-20 | Numat, Inc. | Pulsed neutron elemental on-line material analyzer |
| KR20010016730A (en) * | 1999-08-03 | 2001-03-05 | 이계호 | Preparation of standard solution using neutron active analysis |
| US7064337B2 (en) * | 2002-11-19 | 2006-06-20 | The Regents Of The University Of California | Radiation detection system for portable gamma-ray spectroscopy |
| US20050254614A1 (en) * | 2004-03-11 | 2005-11-17 | Mckinny Kevin S | Method and apparatus for measuring wall thickness of a vessel |
| US8338793B2 (en) * | 2005-04-28 | 2012-12-25 | Utah State University | Identification and localization of explosives and other material |
| ES2327990B1 (en) * | 2006-05-03 | 2010-08-30 | Universidad De Oviedo | METHOD AND APPLIANCE FOR FLUOR ANALYSIS, FROM MINERAL SAMPLES OR FLUOR COMPOUNDS, BY THE METHOD OF NEUTRONIC ACTIVATION. |
| US9389191B2 (en) * | 2012-10-22 | 2016-07-12 | Troxler Electronic Laboratories, Inc. | Conveyor system and measuring device for determining water content of a construction material |
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| RU2648105C1 (en) * | 2017-02-09 | 2018-03-22 | Общество с ограниченной ответственностью "Диамант" | Separator and method of dry concentration of diamond-containing ore |
| CN108344757A (en) * | 2018-01-26 | 2018-07-31 | 吉林大学 | The device of talcum powder content in a kind of detection flour |
| CN108051465A (en) * | 2018-01-26 | 2018-05-18 | 吉林大学 | A kind of Atomic Absorption SpectrophotometerICP neutron activation wink sent out gamma ray deferred and combined |
| CN112432965A (en) * | 2019-08-10 | 2021-03-02 | 丹东东方测控技术股份有限公司 | Method for on-line detection of sinter components |
| JP2022157653A (en) * | 2021-03-31 | 2022-10-14 | 株式会社トプコン | Non-destructive inspection device |
Family Cites Families (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3008047A (en) * | 1956-07-23 | 1961-11-07 | Gulf Research Development Co | Induced gamma activity prospecting method |
| US3035174A (en) * | 1958-07-28 | 1962-05-15 | Socony Mobil Oil Co Inc | Silica-alumina radioactive logging |
| US3463922A (en) * | 1966-07-29 | 1969-08-26 | Us Interior | Mineral ore exploration apparatus utilizing neutron activation |
| US3665195A (en) * | 1969-10-16 | 1972-05-23 | Dresser Ind | Thermal neutron activation radioactivity logging method |
| US3781545A (en) * | 1971-05-26 | 1973-12-25 | Texaco Inc | Well logging comprising detecting unstable radioactive isotopes of aluminum |
| AU475297B2 (en) * | 1972-06-09 | 1976-08-19 | Commonwealth Scientific And Industrial Research Organisation | Analysis utilizing neutron irradiation |
| US3781556A (en) * | 1972-09-08 | 1973-12-25 | Atomic Energy Commission | Neutron activation analysis system |
| ZA766086B (en) * | 1975-10-29 | 1977-07-27 | Atomic Energy Commission | Analysis of coal |
-
1980
- 1980-08-04 FI FI802432A patent/FI73527C/en not_active IP Right Cessation
- 1980-08-05 DE DE8080302680T patent/DE3068109D1/en not_active Expired
- 1980-08-05 EP EP80302680A patent/EP0024835B1/en not_active Expired
- 1980-08-06 US US06/175,799 patent/US4361534A/en not_active Expired - Lifetime
- 1980-08-06 CA CA000357743A patent/CA1156377A/en not_active Expired
- 1980-08-06 ZA ZA00804773A patent/ZA804773B/en unknown
Also Published As
| Publication number | Publication date |
|---|---|
| EP0024835B1 (en) | 1984-06-06 |
| EP0024835A1 (en) | 1981-03-11 |
| DE3068109D1 (en) | 1984-07-12 |
| US4361534A (en) | 1982-11-30 |
| ZA804773B (en) | 1981-11-25 |
| FI73527C (en) | 1987-10-09 |
| FI73527B (en) | 1987-06-30 |
| FI802432A (en) | 1981-02-07 |
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